The present invention relates to a fast reactor, and more particularly, to a fast reactor having a reflector control system for controlling a reactivity of a core by utilizing a neutron reflector.
One example of conventional fast breeder reactor is shown in FIG. 47. Referring to FIG. 47, a fast breeder reactor 10 is provided with a columnar core 11 which is supported by a core barrel 12 disposed outside the core 11 and a reactor vessel 13 is disposed further outside the core barrel 12. A guard vessel 14 for protecting the reactor vessel 13 is disposed outside the reactor vessel 13 and a reflector 15 is disposed further outside the guard vessel 14. A coolant passage 16 through which a primary coolant flows downward is formed between the core barrel 12 and the reactor vessel 13. An electromagnetic pump 17 is disposed perpendicularly above the core, and an intermediate heat exchanger 18 and a decay heat removal coil 19 are disposed further above the electromagnetic pump 17.
In the actual operation of the fast breeder reactor 10 of FIG. 47, the primary coolant such as liquid sodium fills the reactor pressure vessel 13 and plutonium in the core is then fissioned. This core 11 contains plutonium and depleted uranium, and heat is generated in accordance with the fission of the plutonium, thereby emitting neutrons. The emitted neutrons are reflected by the reflector disposed so as to surround the outer periphery of the guard vessel 14 and are then absorbed by the depleted uranium to thereby produce plutonium. The thus produced plutonium is again fissioned and the heat is generated.
In accordance with burn-up of the core, the reflector 15 is relatively vertically moved while maintaining a critical state of the core 11, whereby the burn-up gradually progresses and generates the heat for a long time.
The primary coolant moves upward in the reactor vessel 13 as shown by solid arrow in FIG. 47 by the actuation of the electromagnetic pump 17, decends in the coolant passage 16 through the intermediate heat exchanger 18 and then again flows in the electromagentic pump 17 through the core 11. The primary coolant passes the core 11 while absorbing the heat generated in the core 11 and the heat is transferred to the intermediate heat exchanger 18. A secondary coolant flows into the intermediate heat exchanger 18 through an inlet tube 20 as shown by broken arrow in FIG. 47 and, in the intermediate heat exchanger 18, the heat exchanging operation is carried out between the primary coolant and the secondary coolant. The heat from the core 11 is taken outside the reactor vessel 13 through an outlet tube 21, which is then utilized as a power source.
However, in the conventional fast breeder reactor 10 of the structure shown in FIG. 47, since there is not provided a neutron shield in the reactor vessel and the reflector is disposed outside the reactor vessel, the reactor vessel and the reflector diffuse a large amount of heat inside a shielding structure accommodating the fast breeder reactor. In order to remove this heat, the shielding structure of the conventional fast breeder reactor must be provided with a cooling equipment having large capacity, thus providing a significant problem.
Furthermore, since the conventional fast breeder reactor radiates a large amount of neutrons outside the reactor vessel and gas such as argon and nitrogen in an atmosphere in the shielding structure is activated, it is necessary to provide the activated gas containment vessel for preventing the gas from discharging externally in an environment under severe management, resulting in further enlargement of an entire reactor arrangement, thus also providing a problem.
Still furthermore, in the conventional fast breeder reactor, since a neutron irradiation amount to the reactor vessel during a life time of the reactor exceeds 10.sup.23 nvt (E&gt;0, 1 MeV), stainless steel is not used and expensive crominium steel is to be used, thus also providing an economical problem.
Still furthermore, in the conventional fast breeder reactor, since the electromagnetic pump is disposed directly above the core, a large thermal strain is caused to the electromagnetic pump by the heat of the liquid sodium highly heated by the core and the life time for maintaining required reliability is then shortened, and accordingly, in the conventional fast breeder reactor, the shortening of the life time of the electromagnetic pump adversely affects the life time of a small sized fast breeder reactor itself.
Still furthermore, in the conventional fast breeder reactor, since the intermediate heat exchanger as well as the electromagnetic pump is disposed directly above the core, it is necessary to disassemble and remove the electromagnetic pump and the intermediate heat exchanger at a fuel exchanging time, resulting in a complicated and troublesome disassembling and removing working and a possibility of giving accidental damage to these elements is also increased.
Still furthermore, in the conventional fast reactor having a reflector moving structure, in order to enhance a controlling ability of the neutron reflector, it is obliged to elongate the length of the neutron reflector itself. However, the elongation of the neutron reflector increases its weight, and moreover, affects the core structure itself, and accordingly, it is not desired to elongate the length of the neutron reflector in various view points. Particularly, in so-called a incore reflector type fast reactor in which the neutron reflector is arranged in the reactor vessel, it is difficult to use an elongated neutron reflector from the view point of the incore structure, thus remarkably providing the above problem.
FIG. 48 is an illustration showing a structure of a conventional nuclear power plant 30 including a control system therefor. Referring to FIG. 48, a core 32 is accommodated in a reactor 31 and the core 32 generates heat through a fission chain reaction and heats a primary coolant passing the core. The heated primary coolant is fed into an intermediate heat exchanger 34 through a primary coolant high temperature side line 33 and, in the intermediate heat exchanger 34, heat exchanging operation is performed between the primary coolant and a secondary coolant to transfer the heat to the secondary coolant. After the heat exchanging operation, the primary coolant having the lowered temperature is again circulated into the reactor 31 through a primary coolant low temperature side line 35. Such circulation of the primary coolant is carried out by means of a coolant circulation pump 36.
The secondary coolant having a raised temperature through the heat exchanging operation is transferred to a steam generator 38 as a load heat exchanger through a secondary coolant high temperature side line 37 and heats a water in the steam generator 38. The secondary coolant having temperature lowered in the steam generator 38 is circulated into the intermediate heat exchanger 34 through a secondary coolant low temperature side line 39. Such circulation of the secondary coolant is performed by means of a secondary coolant circulation pump 40.
The water heated through the heat exchanging operation in the steam generator 38 changes to a steam, which is fed to a turbine 42 and drives the same to thereby generate power. The water is fed to the steam generator 38 by means of a water feed pump 43 through a water feed line 57 and feed water flow rate Gw is regulated by a feed water flow rate regulating valve 44.
Power control in the conventional nuclear power plant 30 is performed in the following manner.
The control system of the nuclear power plant 30 comprises a power setter 45 for setting a power, a reactor power control unit 47 for controlling a control rod 46, a primary coolant flow rate regulator 48 for regulating the flow rate of the primary coolant, a secondary coolant flow rate ragulator 49 for regulating the flow rate of the secondary coolant, and a feed water flow rate regulator 50 for regulating the feed water flow rate Gw to the steam generator 38.
The reactor power control unit 47 operates and processes a driving speed of the control rod in response to a power setting signal from the power setter 45, with a reactor outlet temperature detected by a temperature detector 51 being as a feddback signal and a neutron flux level detected by the neutron detector 51 being an auxiliary signal, and then controls the vertical movement of insertion or withdrawal of the control rod 46 in accordance with the operated and processed result. The power of the reactor 31 is regulated by vertically moving the control rod 46.
The primary coolant flow rate regulator 48 controls the revolution number of the primary coolant circulation pump 36 in response to the power setting signal form the power setter 45 with the flow rate of the primary coolant detected by the primary coolant flow rate detector 53 being a feedback signal. The flow rate of the primary coolant is regulated by changing the revolution number of the primary coolant circulation pump 36.
The secondary coolant flow rate regulator 49 controls the revolution number of the secondary coolant circulation pump 40 in response to the power setting signal from the power setter 45 with the flow rate of the secondary coolant detected by the secondary coolant flow rate detector 54 being as a feedback signal. The flow rate of the primary coolant is regulated by changing the revolution number of the primary coolant circulation pump 40.
The feed water flow rate regulator 50 controls an opening degree of the feed water regulating valve 44 in response to the power setting signal from the power setter 45 with the feed water flow rate detected by the feed water flow rate detector 55 being as a feedback signal and a steam temperature detected by the steam temperature detector 56 being as an auxiliary signal. The feed water flow rate to the steam generator 38 is regulated by changing the opening degree of the feed water flow rate regulating valve.
As described above, in the conventional nuclear power plant 30, the inserting, i.e. charging, amount or degree of the control rod 46, the flow rates of the primary and secondary coolants and the feed water flow rate to the steam generator 38 are set by the power setter 45, and in order to maintain the set values regarding these factors, the power setter 45, the reactor power control unit 47, the primary coolant flow rate regulator 48, the secondary coolant flow rate regulator 49 and the feed water flow rate regulator 50 are operated, thereby maintaining the value of the aimed power.
However, the control system of the conventional nuclear power plant is composed of the power setter, the reactor power control unit, the primary coolant flow rate regulator, the secondary coolant flow rate regulator and the feed water flow rate regulator, thus being complicated in its structure. Furthermore, since the reactor power control unit directly operates the control rod, there is a fear of erroneously withdrawing the control rod due to a failure of the reactor power control unit itself. This problem has been commonly considered to the case of a fast breeder reactor in which the power is roughly adjusted by driving the reflector and a fear resides in an erroneous operation of the reflector.